1. Field of the Invention
This invention relates to a continuous process electrorefiner, and more specifically, this invention relates to an improved continuous process electrorefiner for recycling components of spent metallic nuclear fuel, such as uranium.
2. Background of the Invention
Uranium is the naturally-occurring material upon which conventional nuclear power is based. When the fissile uranium-235 isotope absorbs a neutron, fission occurs, with the liberation on average, of approximately 2.5 neutrons. Some of these neutrons are used to bombard more uranium, while other of these neutrons are used to create plutonium (Pu) by the reaction:238U+1n→239U→239Np→239Puand subsequently fission some of it. The energy of fission fragments is used to heat water, gas, or liquid metal. These heated fluids in turn are used to spin electric-generating turbines.
Uranium is scattered in deposits throughout the world. Further, its total supply is not known. The efficiency of use of the energy locked up in uranium can be very low. Approximately one to two percent of the energy content of uranium is tapped in uranium-235-based nuclear power systems. The remaining 98 to 99% of the energy content of uranium is present as uranium-238 which can be converted into fissionable plutonium-239 via neutron bombardment in breeder reactors. Otherwise, “spent” metallic uranium fuel, i.e., having little uranium-235, and the bulkiness of the materials associated with that “spent” fuel present storage and disposal problems.
Current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for spent nuclear fuel in the U.S., and driving up its cost. The resistance among Yucca Mountain Range residents and others regarding plans to deposit radioactive material in the Yucca Mountain Repository is a case in point.
Instead of long term storage of untreated radioactive materials, preliminary treat-ment of spent nuclear fuel is being explored, including partial utilization of the fissile material contained in the spent fuel via conversion to plutonium-239 in breeder reactors. Accordingly, there is an emphasis upon developing new technologies for reprocessing and reutilizing spent nuclear fuels.
A number of processes exist for the processing and recycling of nuclear fuels. These processes often involve aqueous solutions. Due to the presence of water, aqueous solutions are neutron moderators. This is because collisions between water nuclei and neutrons, which are initially created by the spontaneous fission of plutonium, lowers the neutrons' kinetic energies. This lower energy increases the likelihood of the neutrons inducing more fission upon their collision with the plutonium nuclei remaining in the fuel. Thus, previously innocuous levels of plutonium now become potential run-away fission hazards. This lowered critical mass necessitates the use of very low plutonium concentrations and redundant safeguards to assure fission control. Lower plutonium throughputs result. Aqueous solution processing and recycling of nuclear fuels is generally inefficient and not cost-effective.
Research in pyrometallurgical processing of spent metallic nuclear fuel continues for reducing both the radiotoxicity and the volume of waste from commercial nuclear power generation. This is especially true when such pyrometallurgical processing is combined with either a reactor-based or accelerator-driven actinide burner. A major step in this process is the electrorefining and separation of uranium, the major heavy metal component of the spent fuel, from the higher actinides such as plutonium, so that the latter may be fabricated into new fuel assemblies for insertion into the reactor-driven or accelerator-driven burner.
Interest in recent years has focused on electrorefining the large inventories of blanket fuel and other spent metal fuels at DOE sites such as the approximately 25 metric tons stored at the Experimental Breeder Reactor Two (EBR-II) at Argonne National Laboratory-West in Idaho. (Blanket fuel contains primarily uranium-238, a non-fissile isotope that a reactor converts to fissionable plutonium. The blanket fuel is encased in steel cladding and is situated beyond the reactor core's outer edge and thus forms a “blanket” around the core. The name has also been given to similar assemblies located within the plutonium-fueled driver core.)
A current method in the art for electrorefinement of these spent metallic nuclear fuels is a circular batch processor. Generally, these processors have a small throughput.
The typical electrorefiner consists of a hollow cathode, about 10 inches in diameter and about 10 inches in height. Anode baskets are attached to a central spindle and designed to rotate coaxially and within the center of the cathode. Several of these anode/cathode assemblies or modules are located within a larger container of electrolytic salt.
FIG. 1 depicts a cross-sectional view of such a design, designated generally as numeral 10. In this design 10, depleted nuclear fuel is loaded into anode baskets (all of approximately the same volume) 12, made of ferrous metals, that rotate in two channels 14. Each anode basket 12 has a bus bar 16 which serves as a metal spine for the anode basket 12. The baskets are positioned between cylindrical cathode tubes 18. The anode basket assembly 12 is attached to a circular plate and a central spindle (not shown) which are located above the baskets. Current flows from the circular plate to the bus bars 16 which distribute the current to and through the anode baskets 12. The anode assembly 12 and the cathode tubes 18 are submersed in a molten LiCl—KCl eutectic (not shown) which is situated in the channels 14. The salt also contains 2 to 3 mole (mol) % uranium as uranium (III) U3+ cations, provided by adding UCl3 to the salt.
Uranium and the elements in the fuel that are less noble than uranium are oxidized at the anode baskets 12 (U0 to U+3) and form cationic species that dissolve in the molten salt. Zirconium (Zr) and noble metal fission products, such as molybdenum (Mo), ruthenium (Ru), palladium (Pd), platinum (Pt), and rhodium (Rh), remain in the anode baskets 12 inasmuch as the optimal applied voltage is too low to ionize these metals. Gaseous fission products escape and rare earth metal fission products dissolve in the molten eutectic and remain there. The molten salt is eventually cleaned.
Uranium cations, liberated at the anode, migrate to and are then reduced by the cathode 18 and deposited thereon. Scrapers 20 which form part of the bus bars 16 dislodge the electrodeposited uranium, which then falls into a collection basket (not shown) positioned inferior to a depending or bottom region of the outer cathode tube 18. Since the scrapers 20 are immersed in the eutectic salt bath and contact the anode 12, cathode 18, and the bus bar 16, they are usually made of an insulating ceramic material such as beryllia (beryllium oxide, BeO).
This system 10 experiences frequent binding and consequent stalling of the rotation drives, partially due to the buildup of a lumpy product containing residual unrefined salt. This salt must be boiled off in a Cathode Processor. These binding problems are due to holdup of material in the narrow annular regions between the rotating segmented cylindrical anodes 12 and surrounding cathodes 18 of the electrode modules. These narrow annular regions cannot be widened, otherwise, the efficiency-robbing resistance will increase between the anode and cathode. As a result of this state of the art design, only a small throughput of metal product is realized.
This typical system 10 requires a high anode 12 surface-to-volume ratio so as to maximize electrochemical efficiency. This requirement inherently constrains the vertical anode fuel-bed baskets 12 to small sizes since they can be enlarged easily only in the vertical length-wise, or load and unload, direction.
Product collection is done at the bottom of the electrode assembly and allows for undesirable contaminants such as fission product particulates to fall into a collection basket along with the uranium product.
At the present time, no continuous and efficient, high throughput process exists for the processing and treatment of metallic spent nuclear fuels.
U.S. Pat. No. 5,650,053 awarded to Gay, et al. on Jul. 22, 1997 discloses a process and device for the electrorefining of spent metallic nuclear fuels. The device has parallel electrodes with anode baskets rotating within a cylindrical cathode.
U.S. Pat. Nos. 5,531,868 and 5,443,705 awarded to Miller, et al. on Jul. 2, 1996, and on Aug. 22, 1995, respectively, disclose processes and devices for the electrorefining of uranium and plutonium.
U.S. Pat. No. 5,372,794 awarded to LeMaire, et al. on Dec. 13, 1994 discloses a process for separation of actinides from aqueous solutions.
U.S. Pat. No. 5,132,092 awarded to Musikas on Jul. 21, 1992 discloses an aqueous process for the extraction of uranium (VI) and plutonium (IV).
U.S. Pat. No. 5,085,834 awarded to LeMaire, et al. on Feb. 4, 1992 discloses an aqueous method for separating plutonium from uranium and from fission products.
U.S. Pat. No. 5,009,752 awarded to Tomczuk, et al. on Apr. 23, 1991 discloses a process and device for the electrorefinement of spent metallic nuclear fuel in the form of steel-clad metal pins containing 90% uranium and 10% zirconium (Zr).
U.S. Pat. No. 4,740,359 awarded to Hadi Ali, et al. on Apr. 26, 1988 discloses an organic-aqueous process for recovering uranium values.
U.S. Pat. No. 4,399,108 awarded to Krikorian et al. on Aug. 16, 1983 discloses a carbothermic reduction method for the recovery of actinides.
U.S. Pat. No. 4,297,174 awarded to Brambilla on Oct. 27, 1981 discloses a pyroelectrochemical process for reprocessing irradiated nuclear fuels. The process involves dissolving fuel to be reprocessed in a fused-salt bath.
U.S. Pat. No. 4,092,397 awarded to Brambilla, et al. on May 30, 1978 discloses a method for the pyrochemical separation of plutonium from irradiated nuclear fuels, by thermal decomposition in molten nitrates.
U.S. Pat. No. 3,981,960 awarded to Brambilla, et al. on Sep. 21, 1976 discloses a reprocessing method of ceramic nuclear fuel in low-melting nitrate molten salts.
Several of these patents teach aqueous separation processes which are less than efficient. Other patents amongst these do not disclose a method for the electrorefining of metallic nuclear fuels. Also, none of the aforementioned patents disclose either a process or apparatus to counter the aforementioned difficulties, including the low throughput of refined uranium metal. Further, none of the aforementioned patents anticipate or suggest continuous, uninterrupted electrochemical oxidation and reduction of uranium.
A need exists in the art for an improved method and device for isolating uranium metal from elements of spent metallic nuclear fuel. The method should not require aqueous separation techniques. The method and device should directly and continuously separate uranium metal from the transuranics and noble metal fission products present in the spent metallic fuel. In addition, the method and device should have a much higher throughput of refined uranium metal than present uranium electrorefiners.